A Numerical Characterization of Flow-Induced Vibration and Fretting Wear Potential in Nuclear Steam Generator Tube Bundles

2011 ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Flow Induced Vibration ◽  
Numerical Characterization ◽  
Wear Damage ◽  
Effect Of Support ◽  
Tube Failure

Nuclear power plants have experienced problems related to tube failures in steam generators. While many of these failures have been attributed to corrosion, it has been recognized that flow-induced vibrations contribute significantly to tube failure. In order to avoid these excessive vibrations, tubes are stiffened by placing supports along their length. Various tube/support geometries have been used, but the majority are either support plates (plates with drilled or broached holes) or flat bars. Unfortunately, clearance is often considered necessary between the tubes and their supports to facilitate tube/support assembly and to allow for thermal expansion of the tubes. A combination of flow-induced turbulence and fluidelastic forces may then lead to unacceptable tube fretting-wear at the supports. The fretting wear damage could ultimately cause tube failure. Such failures may require shut downs resulting in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting wear damage to tubes due to fluid excitations. Tubes in loose flat-bar supports have complex dynamics due to the possible combinations of geometry. The understanding of tube dynamics in the presence of this type of support and the associated fretting wear is still incomplete. These issues are addressed in this paper through simulations of the dynamics of tubes subjected to crossflow turbulence and fluidelastic instability forces. The finite element method is utilized to model the vibrations and impact dynamics. The tube model simulates a U-tube supported by 16 flat bars with clearances and axial offset. Results are presented and comparisons are made for the parameters influencing the fretting-wear damage such as contact ratio, impact forces and normal work rate. The effect of support clearance and support axial offset are investigated.

Fluidelastic Instability Modeling of Loosely Supported Multispan U-Tubes in Nuclear Steam Generators

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Tube Bundle ◽  
Design Guidelines ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Steam Generators ◽  
Fluidelastic Instability ◽  
Wear Damage

Steam generators in nuclear power plants have experienced tube failures caused by flow-induced vibrations. Two excitation mechanisms are responsible for such failures; random turbulence excitation and fluidelastic instability. The random turbulence excitation mechanism results in long-term failures due to fretting-wear damage at the tube supports, while fluidelastic instability results in short-term failures due to excessive vibration of the tubes. Such failures may require shutdowns, which result in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting-wear damage to tubes due to fluid excitation. In this paper, a numerical model is developed to predict the nonlinear dynamic response of a steam generator with multispan U-tubes and anti-vibration bar supports, and the associated fretting wear due to fluid excitation. Both the crossflow turbulence and fluidelastic instability forces are considered in this model. The finite element method is utilized to model the vibrations and impact dynamics. The tube bundle geometry is similar to the geometry used in CANDU steam generators. Eight sets of flat-bar supports are considered. Moreover, the effect of clearances between the tubes and their supports, and axial offset between the supports are investigated. The results are presented and comparisons are made for the parameters influencing the fretting-wear damage, such as contact ratio, impact forces, and normal work rate. It is clear that tubes in loose flat-bar supports have complex dynamics due to a combination of geometry, tube-to-support clearance, offset, and misalignment. However, the results of the numerical simulation along with the developed model provide new insight into the flow-induced vibration mechanism and fretting wear of multispan U-tubes that can be incorporated into future design guidelines for steam generators and large heat exchangers.

Fretting Wear Damage of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Grids

2007 ◽  
Vol 345-346 ◽  
pp. 709-712
Author(s):  
Jin Seon Kim ◽  
Yong Hwan Kim ◽  
Seung Jae Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Fluid Flows ◽  
Fretting Wear ◽  
Wear Depth ◽  
Fuel Cladding ◽  
Flow Induced Vibration ◽  
Nuclear Fuel Assembly ◽  
Wear Damage ◽  
Increasing Load

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

The Effect of Coating on the Fretting Wear Between Tubes and Supports in a Nuclear Steam Generator

2007 ◽  
Author(s):  
Joong-Hui Lee ◽  
Jin-Sun Kim ◽  
Jung-Min Park ◽  
Bo-Ra Shin ◽  
Young-Ze Lee
Keyword(s):  
Steam Generator ◽  
Coating Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Hard Coatings ◽  
Fretting Wear ◽  
Tube Material ◽  
Flat Type ◽  
Wear Damage ◽  
Wear Tests

The steam generator in the nuclear power plants is a kind of heat exchanger, which is composed of bundles of long and slender pipes. The tubes are supported by anti-vibration bar to reduce the vibration, caused by the water flows for cooling purpose. The wear damage due to this vibration is called as the fretting wear, which should be minimized for the safe operation of plants. The hard coatings are very effective to reduce the amount of wear. In this paper, the coatings of TiN and CrN were deposited on the tube material to protect the fretting surfaces. The tube-on-flat type tester was used for fretting wear tests. The wear amounts of the coated tubes were decreased depending on coating thickness. CrN coating was very effective to reduce the wear. In case of TiN the wear amounts were dependent on the coating thickness. Thick coating of TiN was very effective for wear resistance.

Wear Transitions of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions

2006 ◽  
Vol 326-328 ◽  
pp. 1263-1266 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Jung Min Park ◽  
Joong Hui Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Applied Load ◽  
Fluid Flows ◽  
Relative Displacement ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Support Materials ◽  
Mild Wear ◽  
Inconel 690

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on fretting wear transitions from mild wear to severe wear of tube-support materials by various loads and relative displacements. The transition is defined on the basis of the changes in wear amount. To investigate the transition, the fretting wear tester was contrived to prevent the reduction of relative displacement between tube and support by increasing the load. The tube and support materials were Inconel 690 and 409 SS, respectively. The results show that the transition of tube-support wear is caused by the changes of the dominant wear mechanism depending on the applied load and the relative displacement.

Wear Characteristics of INCONEL 690 and INCONEL 600 in Elevated Temperature

2005 ◽  
Vol 297-300 ◽  
pp. 1424-1429 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Young Ze Lee
Keyword(s):  
Wear Mechanisms ◽  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Fretting Wear ◽  
Wear Characteristics ◽  
Inconel 600 ◽  
Wear Life ◽  
Inconel 690 ◽  
C 50

In this paper, the fretting wear characteristics of INCONEL 690 (I-690) and INCONEL 600 (I-600) was evaluated to verify the wear mechanism and the wear life. Because of the excellent corrosion-resistance of nickel-based alloy, those materials are used for steam generator tube in nuclear power plants. Sometimes the tubes are damaged due to small amplitude vibration, so called fretting wear. To verify the fretting wear mechanisms the wear experiment was carried with the crossed-cylinder wear tester, which used a cam to oscillate the specimen. The test was carried out at loads of 40N and 90N in elevated temperatures of water. The temperatures of water were 20°C, 50°C and 80°C. The increase of water temperature causes the oxidation of the contact area to be delayed, and the amount of wear on oxide layer to be reduced. The main wear mechanisms of fretting were abrasive wear and oxidation wear.

Direct Numerical Simulation of a Turbulent Flow on an Oscillating Wall Using Cut-Cell Based Immersed Boundary Method

2016 ◽  
Author(s):  
Chiaki Kino
Keyword(s):  
Turbulent Flow ◽  
Nuclear Power ◽  
Power Plants ◽  
Separated Flow ◽  
Immersed Boundary ◽  
Turbulent Structures ◽  
Flow Induced Vibration ◽  
Production Term ◽  
Large Eddy Simulation Model ◽  
Wide Range

The flow-induced vibration of a pipe is an important issue in various engineering fields, and this phenomenon is widely observed in nuclear power plants. Although turbulent structures play important roles in the velocity and pressure fields in a pipe, only a few studies have been conducted on the turbulent flow on an oscillating wall. In this study, direct numerical simulations were conducted to establish a large eddy simulation model for a turbulent flow on an oscillating wall and scrutinize the energy transfer between the grid scale (GS) and sub-grid scale (SGS). Although energy is generally transferred from the GS to SGS (forward scatter), it is likely that energy is transferred from the SGS to GS (backward scatter) under specific conditions. The present numerical results indicated that backward scatter exists in the production term in the case of a static wavy wall. On the other hand, such backward scatter could not be observed in the case of an oscillating wall. It is well known that separated flows and backward flows are generated behind the crest. Stronger backward flows accelerate the main flow and enhance the velocity gradients in a wide range behind the crest. In the case of an oscillating wall, the development of separated flow is immature because the shape of the wall is not fixed. Eventually, the backward scatter is deemed to be suppressed.

Design and Maintenance in an IAEA Technical Guidelines on Fluid-Structure Interaction

2014 ◽  
Author(s):  
Fumio Inada ◽  
Tomomichi Nakamura ◽  
Takashi Nishihara ◽  
Shigehiko Kaneko ◽  
Manwoong Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
General Description ◽  
Fluid Structure Interactions ◽  
Flow Induced Vibration ◽  
Energy Agency ◽  
Fluid Structure ◽  
Component Systems

In nuclear power plants, fluid structure interactions (FSI) occurring in component systems can cause excessive forces or stresses to the structures resulting in mechanical damages that may eventually threaten the structural integrity. FSI in the guidelines includes flow-induced vibration, water hammer, and pipewhip. It can also include movement, deformation, or fracture of equipments by tsunami etc. They can be issues of design and maintenance. Authors cannot find complete guidelines to correspond to the FSI phenomena which can be important in the design and maintenance of nuclear power plants. Based on the background, International Atomic Energy Agency (IAEA) has drafted guidelines on FSI. This paper summarizes general description of FSI as well as design and maintenance against FSI.

Experimental Fretting-Wear Studies of Steam Generator Materials

1995 ◽  
Vol 117 (4) ◽  
pp. 312-320 ◽  
Author(s):  
N. J. Fisher ◽  
A. B. Chow ◽  
M. K. Weckwerth
Keyword(s):  
Steam Generator ◽  
Contact Force ◽  
Environmental Conditions ◽  
Analytical Techniques ◽  
Work Rate ◽  
Fretting Wear ◽  
Flow Induced Vibration ◽  
Dynamic Interactions ◽  
Support Materials ◽  
Wear Damage

Flow-induced vibration of steam generator tubes results in fretting-wear damage due to impacting and rubbing of the tubes against their supports. This damage can be predicted by computing tube response to flow-induced excitation forces using analytical techniques, and then relating this response to resultant wear damage using experimentally derived wear coefficients. Fretting-wear of steam generator materials has been studied experimentally at Chalk River Laboratories for two decades. Tests are conducted in machines that simulate steam generator environmental conditions and tube-to-support dynamic interactions. Different tube and support materials, tube-to-support clearances, and tube support geometries have been studied. The effect of environmental conditions, such as temperature, oxygen content, pH and chemistry control additive, have been investigated as well. Early studies showed that damage was related to contact force as long as other parameters, such as geometry and motion, were held constant. Later studies have shown that damage is related to a parameter called work-rate, which combines both contact force and sliding distance. Results of short and long-term fretting-wear tests for CANDU steam generator materials at realistic environmental conditions are presented. These results demonstrate that work-rate is an appropriate correlating parameter for impact-sliding interaction.

Risks for Flow Induced Vibrations (FIV) at EPU

2008 ◽  
Author(s):  
Per Nilsson ◽  
Eric Lillberg
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cross Flow ◽  
Flow Induced Vibration ◽  
Resonance Vibrations ◽  
Flow Induced Vibrations ◽  
Risk Areas ◽  
Tube Banks ◽  
Semi Empirical ◽  
Simple Search

This work deals with risk areas for flow induced vibration at extended power uprate, EPU. The focus is on the mechanisms of excitation in one phase relevant for Swedish BWRs and PWRs. FIV-events that have occurred in nuclear power plants over the world have been collected and categorized. The most relevant events for EPU are summarized to: vibrations in steam systems due to turbulence or vortex shedding and resonance, vibrations of internal parts and also thermal mixers and sleeves or in valves and vibrations of tube banks in partial or full cross flow. Based on the collected events and some semi-empirical methods, a simple search list for FIV by power uprate has been developed. In principle these changes lead to increased risks: changed flow velocity, decreased water temperature and increased steam temperature and decreased structural damping, mass or stiffness. In addition to that, the typical collected events should be regarded.

309 High Temperature Impact-Fretting Wear Experiments for Evaluation of Pipe Wall Thinning Phenomena in Nuclear Power Plants

2013 ◽  
Vol 2013.49 (0) ◽  
pp. 87-88
Author(s):  
Yoshiki SATO ◽  
Akira IWABUCHI ◽  
Michimasa UCHIDATE ◽  
Hitoshi YASHIRO ◽  
Akito OYAKAMA ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pipe Wall ◽  
Fretting Wear ◽  
Wall Thinning ◽  
Temperature Impact
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